The following relates to the nuclear reactor arts, electrical power generation arts, nuclear reactor control arts, nuclear electrical power generation control arts, thermal management arts, and related arts.
In nuclear reactor designs of the pressurized water reactor (PWR) type, a radioactive reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel. The primary coolant is maintained in a compressed or subcooled liquid phase and is discharged out of the pressure vessel into an external steam generator, or alternatively an internal steam generator is located within the pressure vessel (sometimes called an “integral PWR” design). In either design, heated primary coolant water heats secondary coolant water in the steam generator to generate steam. An advantage of the PWR design is that the steam comprises secondary coolant water that is not exposed to the radioactive reactor core.
In a typical integral PWR design configuration, the primary coolant flow circuit is defined by a cylindrical pressure vessel mounted generally upright (that is, with the cylinder axis oriented vertically) and a hollow cylindrical central riser disposed concentrically inside the pressure vessel. Primary coolant flows upward through the reactor core where it is heated and rises through the central riser, discharges from the top of the central riser and reverses direction to flow downward back to the reactor core through a downcomer annulus defined between the pressure vessel and the central riser. This is a natural convection flow circuit that can be powered by heating caused by the reactor core and cooling of the primary coolant as it flows upward and away from the reactor core. However, for higher power reactors it is advantageous or necessary to supplement or supplant the natural convection with motive force provided by electromechanical reactor coolant pumps.
Most commercial PWR systems employ external steam generators. In such systems, the primary coolant water is pumped by an external pump connected with external piping running between the PWR pressure vessel and the external steam generator. This also provides motive force for circulating the primary coolant water within the pressure vessel, since the pumps drive the entire primary coolant flow circuit including the portion within the pressure vessel.
Fewer commercial “integral” PWR systems employing an internal steam generator have been produced. In existing PWR designs, a reactor coolant pump of the “glandless” type using in boiling water reactor (BWR) designs is adapted for use in the integral PWR. The pump is typically coupled into the pressure vessel at the bottom of the pressure vessel, near the reactor core, analogous to the arrangement in a BWR, or at the top of the vessel. Both arrangements are problematic. Coupling into the bottom of the vessel introduces vessel penetrations at low elevation, which could be problematic in the event of a loss of coolant accident (LOCA) involving these connections. Coupling into the top of the vessel is problematic because that region is typically already occupied by external control rod drive mechanism (CRDM) units and by an internal pressurizer or a welded connection with an external pressurizer, as well as by various instrumentation feedthroughs and so forth.
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